1. Field of the Invention
This invention generally relates to a method and apparatus for testing the susceptibility of a nickel-based alloy to undergo stress corrosion cracking within a high temperature water environment. It is particularly useful in predicting, on a substantially accelerated basis, when the Inconel.RTM. heat exchanger tubes of a nuclear steam generator will crack as a result of intergranular stress corrosion.
2. Description of the Prior Art
Methods and devices for determining the susceptibility of a nickel-based alloy to stress corrosion cracking within a steam generator environment are known in the prior art. Such tests are very useful in nuclear steam generator maintenance, because they provide an indication as to when certain preventative maintenance procedures, such as sleeving, should be initiated within the heat exchanger tubes of such generators. Such maintenance procedures prevent radioactive water from the primary side of the generator from leaking out of cracked heat exchanger tubes and contaminating the non-radioactive water used in the secondary side of the generator which forms the steam which ultimately drives the electric turbines in the plant. However, before the utility of such test methods can be fully appreciated, a more complete understanding of the structure and maintenance of nuclear steam generators is necessary.
Nuclear steam generators are comprised of three principal parts, including a secondary side, a tubesheet in which a plurality of U-shaped tubes are mounted, and a primary side. The tubesheet and U-shaped tubes hydraulically isolate the primary from the secondary sides of the steam generator while thermally connecting them together, so that heat from the radioactive water in the primary side is transferred to the non-radioactive water in the secondary side. This heat transfer is accomplished by the U-shaped tubes mounted in the tubesheet which extend throughout the secondary side of the steam generator. The inlet and outlet ends of these U-shaped tubes are mounted in the side of the tubesheet that faces the primary side of the generator. The primary side in turn includes a divider plate that hydraulically isolates the inlet ends of the U-shaped tubes from the outlet ends. Hot, radioactive water heated by the nuclear reactor is admitted into the section of the primary side containing all the inlet ends of the U-shaped tubes. This hot water flows through these inlets, up through the tubesheet, and circulates within and around the U-shaped tubes that extend within the secondary side of the steam generator. This hot, radioactive water transfers its heat through the walls of the U-shaped tubes to non-radioactive feed water present in the secondary side of the generator, thereby converting this feed water into non-radioactive steam. After the nuclear-heated water circulates through the U-shaped tubes, it flows back through the tubesheet, through the outlets of the U-shaped tubes, and into the outlet section of the primary side, where it is circulated back to the nuclear reactor.
The heat exchanger tubes of such nuclear steam generators can suffer a number of different types of corrosion degradation, including intergranular stress corrosion cracking. In situ examination of the tubes within these generators has revealed that most of this intergranular stress corrosion cracking occurs around the tubesheet region of the generator, where the inlet and outlet ends of the U-shaped tubes extend through bores in the tubesheet. Often there is some amount of annular space between the walls of the tube-receiving bores in the tubesheet and the outer walls of the tubes themselves. Experience has shown that potentially corrosive sludges can accumulate on the upper surface of the tubesheet and collect in these annular spaces over long periods of time. To prevent these potentially corrosive sludges from collecting within these annular spaces from the effect of gravity, the heat exchanger tubes are often radially expanded by means of a mechanical or a hydraulic mandrel in order to minimize the clearance between the outer walls of the tubes and the inner walls of the bores in the tubesheet through which they extend. However, some of these potentially corrosive sludges may still collect in these very small spaces between the tubes and the bores of the tubesheets. Moreover, the relatively poor hydraulic circulation of the water in these regions tends to maintain the sludge in these crevices and to create localized "hot spots" in the tubes adjacent the sludge. The heat radiating from these "hot spots" may assist in the corrosion processes that operate on the exterior surfaces of the heat exchanger tubes in chemical combination with corrosive species in the sludge. While most nuclear steam generators include blow-down systems for periodically sweeping the sludge out of the generator vessel, the sludges in the annular crevice regions are not easily swept away the hydraulic currents produced by such systems. Despite the fact that the heat exchanger tubes of such nuclear steam generators are typically formed from corrosion-resistant Inconel.RTM., the constant exposure to corrosive sludges and heat, in combination with the mechanical stresses induced in these walls as a result of the mechanical or hydraulic expansion, can ultimately cause the heat exchange tubes to corrode and crack due to intergranular stress corrosion cracking. This, in turn, can allow radioactive water from the primary side of the steam generator to leak into the secondary side of the generator, thereby radioactively contaminating the steam produced by the generator which turns the blades of the electric turbines of the plants.
Such radioactive contamination of the generator steam can be avoided if certain maintenance procedures, such as tube sleeving, are undertaken before the walls of the tubes crack. In such sleeving operations, a reinforcing sleeve is slid up the heat exchanger tubes in the sections of the tubes surrounded by the tubesheet, and rolled and brazed onto the inner walls of the tubes. But, while such sleeving operations are very effective in extending the useful lifetime of the nuclear steam generator, they are also quite expensive. The steam generator has to be completely shut down and taken off-line (which can cost over $500,000 per day), and very specialized tools and procedures must be utilized to install such sleeves within the radioactive environment inside the steam generator. It is therefore desirable that such sleeving operations be undertaken only when stress corrosion cracking is imminent so that the number of times that the generator must be shut-down for an a maintenance operation is kept to a minimum.
The prior art stress-corrosion tests were developed as a result of the need to predict when stress-corrosion cracking was likely to occur in the tubesheet region of a particular nuclear steam generator. Such tests generally involved subjecting a mechanically stressed sample of the particular type of Inconel.RTM. used in the heat exchanger tubes of a particular steam generator to heated water having hydrogen gas dissolved within it. Typically, the water used in the test was heated to a temperature of between 300.degree. C. and 365.degree. C. (or 572.degree. F. to 689.degree. F.). At all times, the water was maintained in a liquid phase by maintaining the atmosphere within the test vessel at a pressure of approximately 15.8 MPa (or 2,200 psi). To simulate the water chemistry inside a nuclear steam generator, substantially pure water with a small amount of hydrogen dissolved therein was used. The radioactive environment of the primary side of such nuclear generators radiolytically creates small amounts of free hydrogen and oxygen gas in this water. To minimize the amount of radiolytically produced oxygen in the generator water, a measured amount of hydrogen gas is deliberately dissolved into it. Such hydrogen curbs the radiolytic production of free oxygen in accordance with Le Chatelier's principle. In steam generators, approximately 25 to 50 cc of hydrogen (as measured at 0.degree. C. and 1 atmosphere absolute pressure) is dissolved into every kilogram of water used in the primary coolant within the steam generator. To simulate this hydrogen component, the partial hydrogen gas pressure in the mixture of pressurized water and hydrogen in the test vessel was initially adjusted to between 250 and 1300 kPa at room temperature.
The foregoing test conditions have been found to produce corrosion cracking in stressed samples of Inconel.RTM. approximately ten times faster than it would take for these samples to exhibit cracking in a real-time mode.
Unfortunately, such prior art testing methods are not without significant shortcomings. Even though such test methods corrode the test samples in an accelerated mode, the exposure times can still be quite long for certain types of heat-treated Inconel.RTM.. For example, for certain types of heat-treated Inconel.RTM. heat exchanger tubes, the exposure time required to produce cracking in even a very highly stressed sample may be as long as 40,000 hours. Secondly, because of its small molecular size, the free hydrogen within the test vessel will tend to slowly diffuse through the seals and walls of the vessel, a phenomenon which necessitates the use of refreshed autoclave systems. Such systems are complex and expensive, and require such components as continuously overpressured make-up tanks, high-pressure pumps, and effluent coolers.
Clearly, there is a need for a testing method and apparatus which is capable of determining, in a greatly accelerated mode, the susceptibility of nickel-based alloys to stress corrosion cracking in simulated steam generator environments. Ideally, such a test method should be able to operate in a predictive mode that is an order of magnitude faster than known stress corrosion cracking tests, while accurately maintaining a calibrated amount of free hydrogen within its test vessel throughout the duration of the test. Finally, it would be desirable if such a vastly accelerated test method could be conducted with a simple and relatively inexpensive test apparatus.